Study On Thermal-Hydraulic Model Of Liquid Lead-Bismuth In Wire Spaced Fuel Assembly

The development level of nuclear energy technology has become an important reflection of the country’s comprehensive strength.Lead cooled fast reactor is a new generation of fast breeder reactor,which uses liquid lead or lead bismuth as coolant and has a good capability of nuclear fuel proliferation and waste disposal.Different from conventional water reactors,the complex flow channel structure of core components and the special thermal and physical properties of liquid metal lead to the different thermal and hydraulic characteristics of coolant in lead cold fast reactor.Therefore,based on the existing experimental data on the flow heat transfer of lead-bismuth medium in wire-spaced fuel pin bundle,this paper verifies the classical thermohydraulic model and numerical simulation method,and constructs a model that can describe the flow heat transfer of lead-bismuth medium in wire-spaced fuel pin bundle.The research contents of this paper can be summarized as follows:Firstly,a frictional pressure drop model is constructed for all flow range.In this paper,the limitations of the existing classical friction pressure drop model are analyzed,the flow range is divided by scientific methods,and a reasonable expression is constructed to describe the factors affecting the flow behavior of fluid in the fuel assembly channel,so as to improve the accuracy of the friction pressure drop model in predicting the experimental results.Finally,the existing friction pressure drop model is analyzed based on the existing experimental data.The results show that the proposed friction pressure drop model has better prediction results in the whole flow state,and the prediction accuracy is significantly improved compared with the existing friction pressure drop model.Secondly,a heat transfer model suitable for different working conditions is proposed.The flow channel structure can affect the heat transfer mechanism of fluid in the flow channel.Based on the analysis of the heat transfer characteristics of the specific flow channel of the fuel assembly under different flow conditions,the heat transfer model under different flow conditions is established.Based on the heat transfer data with low Peclet number,the proposed heat transfer model characterizes the buoyancy effect:the buoyancy changes with the temperature difference between the fluid and the wall,the thickness of the boundary layer changes with the development of the flow state,and the local heat transfer coefficient changes with the thickness of the boundary layer.A heat transfer model based on the inlet effect was constructed to characterize the influence of inadequate development of thermal boundary layer.Due to the thin boundary layer in the inlet section,the local surface heat transfer coefficient of the inlet section was larger than that of the fully developed section,which played a role in strengthening heat transfer.The proposed cross section average heat transfer model is beneficial to grasp the whole situation of the component..Finally,different turbulence models and turbulent Prandtl number models were verified based on the existing experimental data of lead-bismuth medium for widewound fuel assemblies.The SST k-ω model is used to compare and analyze the applicability of different turbulence Prandtl number models,and a turbulence Prandtl number model suitable for specific fuel assembly structure is proposed,which improves the accuracy of calculating low Prandtl number fluids based on Reynolds analog turbulence model.In summary,based on the existing experimental data,some relevant studies are carried out in aspects of analyzing the limitations of classical friction pressure drop model,constructing different heat transfer models,and verifying the applicability of turbulence model and turbulence Prandt number model.The research results can provide certain reference for the design and construction of lead fast reactor in our country.